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PARTICLE code simulationS WITH INJECTED PARTICLES

SPACE SCIENCE REVIEWS 1985年第1-2期42卷 131-144页

作者： ALDRICH, CH Los Alamos National Laboratory Los Alamos New Mexico

As problems we are interested in become more complex, we often find our simulations stretching the limits of available computer resources. For example, an interesting problem is simulation of dissipation processes in sub-critical collisionless shocks. To simulate this system our simulation box must contain the shock and its upstream and downstream regions over the entire length of a run. If the shock moves with any appreciable speed the box must then be considerably larger than the shock thickness making it hard to resolve the shock front itself with a reasonable number of grid points.A solution to this problem is to run the simulation in the frame of reference of the shock. Particles are injected upstream of the shock and leave the simulation box downstream. With the shock stationary in the simulation box, we only need to contain enough of the up and downstream regions for the fields, etc., to settle down and separate the shock from the box *** this tutorial we consider some basic algorithms used in a practical particle injection code, such as the two dimensional WAVE code used at Los Alamos. We will try to present these ideas in a simple format general enough to be easily included in any particle code. Topics covered are: • Smoothly Injecting Particles. • Generating the Distribution Functions. • Time Dependent Injection Density. • Boundary Conditions on Fields and Particles. (Flux and Charge Conservation)

关键词： Shock Front Downstream Region Computer Resource Dissipation Process code simulation

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Main modelling features of the ASTEC V2.1 major version

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ANNALS OF NUCLEAR ENERGY 2016年 93卷 83-93页

作者： Chatelard, P. Belon, S. Bosland, L. Carenini, L. Coindreau, O. Cousin, F. Marchetto, C. Nowack, H. Piar, L. Chailan, L. IRSN PSN RES BP3 F-13115 St Paul Les Durance France Gesell Anlagen & Reaktorsicherheit GRS mbH Schwertnergasse 1 D-50677 Cologne Germany

A new major version of the European severe accident integral code ASTEC, developed by IRSN with some GRS support, was delivered in November 2015 to the ASTEC worldwide community. Main modelling features of this V2.1 version are summarised in this paper. In particular, the in-vessel coupling technique between the reactor coolant system thermal-hydraulics module and the core degradation module has been strongly re-engineered to remove some well-known weaknesses of the former V2.0 series. The V2.1 version also includes new core degradation models specifically addressing BWR and PHWR reactor types, as well as several other physical modelling improvements, notably on reflooding of severely damaged cores, Zircaloy oxidation under air atmosphere, corium coolability during corium concrete interaction and source term evaluation. Moreover, this V2.1 version constitutes the back-bone of the CESAM FP7 project, which final objective is to further improve ASTEC for use in Severe Accident Management analysis of the ***-III nuclear power plants presently under operation or foreseen in near future in Europe. As part of this European project, IRSN efforts to continuously improve both code numerical robustness and computing performances at plant scale as well as users' tools are being intensified. Besides, ASTEC will continue capitalising the whole knowledge on severe accidents phenomenology by progressively keeping physical models at the state of the art through a regular feed-back from the interpretation of the current and future experimental programs performed in the international frame. (C) 2016 Elsevier Ltd. All rights reserved.

关键词： ASTEC Severe accidents code simulation

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SARNET benchmark on Phebus FPT3 integral experiment on core degradation and fission product behaviour

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ANNALS OF NUCLEAR ENERGY 2016年 93卷 65-82页

作者： Di Giuli, M. Haste, T. Biehler, R. Bosland, L. Herranz, L. E. Fontanet, J. Beuzet, E. Torkhani, M. Davidovich, N. Klein-Hessling, W. Weber, S. Dickinson, S. Horvath, G. Kruse, P. Koch, M. Paci, S. Weber, S. J. Salay, M. Bujan, A. Ivanov, I. Kalychev, P. Kim, S. B. Morandi, S. Del Corno, A. Kotouc, M. Dienstbier, J. Kim, H. -C. IRSN Cadarache France CIEMAT E-28040 Madrid Spain EDF Clamart France ENEA Bologna Italy EGRS Garching Germany GRS Cologne Germany NNL Didcot Oxon England NUBIKI Budapest Hungary Ruhr Univ Bochum Bochum Germany Univ Pisa UNIPI I-56100 Pisa Italy SNL Albuquerque NM USA USNRC Washington DC USA VUJE Trnava Slovakia Tech Univ Sofia Sofia Bulgaria KAERI Taejon South Korea RSE Milan Italy UJV Rez Czech Republic KINS Taejon South Korea

The importance of computer simulations in the assessment of nuclear plant safety systems has increased dramatically during the last three decades. The systems of interest include existing or proposed systems that operate, for example, normal operation, in design basis accident conditions, and in severe accident scenario beyond the design basis. The role of computer simulations is especially critical if one is interested in the reliability, robustness, or safety of high consequence systems that cannot be physically tested in a fully representative environment. In the European 7th Framework SARNET project, European Commission (EC) co-funded from 2008 to 2013, the Phebus FPT3 experiment was chosen as a code benchmark exercise to assess the status of the various codes used for severe accident analyses in light water reactors. The aim of the benchmark was to assess the capability of computer codes to model in an integral way the physical processes taking place during a severe accident in a pressurised water reactor (PWR), starting from the initial stages of core degradation, fission product, actinide and structural material release, their transport through the primary circuit up to the behaviour of the released fission products in the containment. The FPT3 benchmark was well supported, with participation from 16 organisations in 11 countries, using 8 different codes. The temperature history of the fuel bundle and the total hydrogen production were well captured. No code was able to reproduce accurately the final bundle state, using as bulk fuel relocation temperature, the temperature of the first significant material relocation observed during the experiment. The total volatile fission product release was well simulated, but the kinetics were generally overestimated. Concerning the modelling of semi-volatile, low-volatile and structural material release, the models need improvement, notably for Mo and Ru for which a substantial difference between bundle and fuel release

关键词： Phebus FPT3 Severe accidents code simulation Benchmark

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Heaving motion effect on the transient thermodynamic characteristics of helical coil once-through steam generators in mixed ocean conditions

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ANNALS OF NUCLEAR ENERGY 2024年 201卷

作者： Zhang, Chao Li, Jiangping Sun, Haoyu Li, Weichao Hou, Yandong Xiang, Yan Northeast Elect Power Univ Sch Energy & Power Engn Jilin Peoples R China Royal Inst Technol KTH Stockholm Sweden

Based on the finite difference method, the fixed boundary method, and a staggered grid method, this paper develops a one-dimensional thermal-hydraulic analysis code to simulate the effects of heave motion and more complex oceanic motions, namely heaving coupled with inclination and heaving coupled with swaying, on the thermal-hydraulic parameters of the International Reactor Innovative and Secure (IRIS) reactor's helical coil once-through steam generator (HCOTSG). The oceanic conditions were achieved by modifying the momentum equation. The validity and accuracy of the code were obtained by comparing the results with the design parameters of HCOTSG for two types of marine nuclear reactors, the SMART reactor and the IRIS reactor, and with the calculations of some similar international prediction programs. The simulation results indicate that (1) when only heave motion is considered, increasing the amplitude and reducing the period both result in a greater magnitude and a faster rate of change in the HCOTSG parameters. (2) When the combined motion of inclining and heaving is present, the dynamic performance of the composite motion inclining towards the positive x-axis deteriorates the most severely, while its thermodynamic conditions improve slightly. In this direction, the inclining angle has little influence on the thermal-hydraulic characteristics of HCOTSG. (3) When the combined motion of heaving and swaying is present, the dynamic performance is more severely affected under swaying motion towards the negative y-axis, despite a slight improvement in thermodynamic conditions. In this direction, as the swaying angle increases, the trend of parameter variation under swaying is similar but with a greater magnitude of change. The deterioration in dynamic performance becomes more severe, although there are small improvements in thermodynamic conditions. With the swaying period increasing, the maximum and minimum values of the secondary side pressure drop show little chan

关键词： Heaving motion Helical coil once -through steam generator code simulation Ocean condition

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Thermal-hydraulic characteristics of helical coiled once-through steam generators in inclining condition of ocean

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ANNALS OF NUCLEAR ENERGY 2024年 200卷

作者： Hou, Yandong Li, Jiangping Sun, Haoyu Li, Weichao Zhang, Chao Xiang, Yan Northeast Elect Power Univ Sch Energy & Power Engn Jilin Peoples R China Royal Inst Technol KTH Stockholm Sweden

This paper develops a one-dimensional thermal-hydraulic analysis program to simulate the main thermal-hydraulic parameter changes of helical coil once-through steam generator (HCOTSG) under inlet thermal-hydraulic parameters perturbation in inclining and vertical conditions. The inclining process is divided into the process of rotating from the vertical position to the inclining position and the stable inclining process. The rotating process is regarded as swaying motion. The swaying and inclining motions are achieved by modifying the momentum equation. What's more, the code was verified through experiments on two-phase heat transfer and friction pressure drop, as well as comparisons with the design parameters of the HCOTSG of the International Reactor Innovative and Secure (IRIS) reactor. The results of the simulation indicate that the direction of incline has an obvious impact on the safety of the HCOTSG, with inclination towards the y-axis having the greatest impact. In the stable inclining process, an increase in the angle of inclining results in a rise in primary-side outlet temperature and a reduction in heat transfer, while the secondary-side pressure drop increases. Furthermore, the HCOTSG underwent testing with eight types of transient perturbations, including four sudden perturbations and four linear perturbations. Results indicated that during sudden perturbations, apart from the secondary-side pressure drop, there were no substantial differences in simulated results between the inclined and vertical states at the same time during the transition process. However, during linear perturbations, due to the slow changes in parameters, the differences between the inclined and vertical states at the same time were more distinct. Regardless of the perturbation type, the inclined state led to a deterioration in the dynamic condition of the system compared to the vertical state.

关键词： Inclination conditions Helical coil once -through steam generator code simulation Transient perturbation

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ANALYSES OF THE PHEBUS FPT3 EXPERIMENT USING THE SEVERE ACCIDENT codeS ATHLET-CD, ICARE/CATHARE, AND MELCOR

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NUCLEAR TECHNOLOGY 2011年第3期176卷 352-371页

作者： Repetto, Georges de Luze, Olivier Drath, Tilman Koch, Marco K. Hollands, Thorsten Trambauer, Klaus Bals, Christine Austregesilo, Henrique Birchley, Jon Inst Radioprotect & Surete Nucl Cadarache St Paul Lez Du France Ruhr Univ Bochum RUB LEE Bochum Germany Gesell Anlagen & Reaktorsicherheit Garching Germany Paul Scherrer Inst Villigen Switzerland

The aim of the Phebus Fission Product (FP) experimental program is to study the degradation phenomena and the behavior of the FPs released in the reactor coolant system and the containment building. The program consists of four in-pile bundle tests (FPTO, FPT1, FPT2, and FPT3), performed under different conditions concerning the thermal hydraulics and the environment of fuel rods, in particular, the amount of steam (strongly or weakly oxidizing atmosphere). The last test of this program, FPT3, was performed in November 2004 in Cadarache. During the FPT3 experiment, for the first time, boron carbide (B4C) was used as the absorber material instead of Ag-In-Cd, which was used in all the previous tests. Boron carbide is used in western-type pressurized water reactors, the EPR, boiling water reactors, and the VVER;consequently, assessing the effects of B4C on the main degradation phenomena and on gas release, as well as its impact on FP behavior is very important. This paper describes results from the Phebus FPT3 experiment, summarizes the test code modeling used in the different code applications, and reports the code results comparing some important experimental parameters, in particular regarding B4C control rod behavior. The severe accident codes used in these studies are Analysis of Thermal-Hydraulics of LEaks and Transients with Core Degradation (ATHLET-CD), ICARE/CATHARE, and MELCOR. The first part is an overview of the experimental results (boundary conditions, temperature evolutions, hydrogen and carbon compound releases coming from the oxidation of the Zircaloy claddings and the B4C absorber, and bundle degradation). The second part summarizes the code modeling used in the different code applications, in particular, those regarding absorber rod degradation and the oxidation process. The third part summarizes the code results comparing some important experimental parameters [thermal behavior, gas releases (H-2, CO, CO2), and bundle degradation]. The conclusion f

关键词： Phebus FP severe accidents code simulation

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Edge plasma modeling of limiter surfaces in a tokamak divertor configuration

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JOURNAL OF NUCLEAR MATERIALS 1999年 266卷 1180-1184页

作者： Rensink, ME Rognlien, TD Univ Calif Lawrence Livermore Natl Lab Livermore CA 94550 USA

We describe a 2-D model for edge plasmas with localized limiter surfaces penetrating the scrape-off layer. The model is used to simulate an idealized ITER startup configuration with variations in the limiter penetration depth and surface shape. The results show that the distribution of the total heat load between divertor plates and outer midplane limiter surfaces can be controlled by the penetration depth of the limiter. The limiter surface can be shaped to reduce the peak heat flux without significantly changing the total heat loads. The tentative conclusion is that ITER startup on an outer midplane limiter is reasonable because the total and peak heat loads can be controlled by the limiter positioning and shape. (C) 1999 Elsevier Science B.V. All rights reserved.

关键词： 2D fluid code 2D modeling boundary plasma code simulation computer simulation divertor divertor modeling divertor plasma edge modeling edge plasma fluid model heat deposition heat flux heat load ITER limiter magnetic surface non-orthogonal grid numerical simulation particle diffusivity particle flow particle flux particle recycling plasma edge plasma edge physics plasma modeling power deposition power flux pumping

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Relentless Repairability or Reckless Reuse Whether or Not to Rebuild a Concern with Your Familiar Tools and Materials

Relentless Repairability or Reckless Reuse Whether or Not to...

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ACM SIGPLAN International Symposium on New Ideas, New Paradigms, and Reflections on Programming and Software (Onward!)

作者： Taeumel, Marcel Hirschfeld, Robert Hasso Plattner Inst Potsdam Germany Univ Potsdam Potsdam Germany

ISBN: (纸本)9781450399098

We must retain liveness and exploratory practices within the programming systems that make us feel most productive. However, the temptation to just reuse black boxes through limited interfaces is pervasive. We expect time savings and better performance at the cost of poor repairability. Fortunately, we also know about the benefits of having an open implementation constructed from familiar materials, integrated with familiar tools. Consequently, it is primarily a matter of "just building it" ... again? Piece of cake. What could possibly go wrong?

关键词： self-sustaining systems exploratory programming liveness reuse open source direct manipulation symbolic debugging code simulation flow Smalltalk Squeak

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Study on the systematic thermal-hydraulic characteristics of helical coil once-through steam generator

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ANNALS OF NUCLEAR ENERGY 2021年 154卷 108096-108096页

作者： Yao, Hao Chen, Guo Lu, Kailin Wu, Yingwei Tian, Wenxi Su, Guanghui Qiu, Suizheng Xi An Jiao Tong Univ Sch Nucl Sci & Technol State Key Lab Multiphase Flow Power Engn Xian 710049 Peoples R China Nucl Power Inst China Chengdu 610226 Peoples R China

Due to the compact structure, the high heat transfer efficiency and the powerful ability to withstand the stress caused by thermal expansion, Helical Coil Once-through Steam Generators (HCOTSGs) have been widely used in recent years, especially as the important equipment of Small-medium Modular Reactors (SMRs). In this study, by adopting the finite difference method, the fixed boundary method, the staggered grid to discretize the control equations and the solution regions, and using the GEAR algorithm to carry out the numerical solution, a systematic Thermal-hydraulic Analysis code for HCOTSG called TACS was developed. Compared with the design values and the calculation values of the ONCESG code of the Marine Reactor X (MRX) steam generator in Japan and the System-integrated Modular Advanced ReacTor (SMART) steam generator in Korea, the correctness of the methods and the models used in the TACS was verified. Based on the TACS, the systematic thermal-hydraulic characteristics of the HCOTSG at full load were studied with a simulation object of the steam generator of the International Reactor Innovative and Secure (IRIS). The results showed that the length of each heat transfer zone is greatly different, so heat flux changes sharply along the tube length. Secondary thermal resistances determine the change of total thermal resistance, decrease at first and suddenly rise when dry-out happens. Frictional pressure drops are the main part of secondary pressure drops, accounting for 90% of the pressure drop in the whole flow process. Besides, the transient processes in HCOTSG with the step and linear disturbances of entrance parameters were simulated by the TACS, which conformed to the analysis of the thermal-hydraulic parameters. The final stable states of these processes are the same under two disturbance modes, but the changing processes caused by linear disturbances are gentler. (C) 2020 Elsevier Ltd. All rights reserved.

关键词： Helical coil once-through steam generator Flow and heat transfer model Thermal-hydraulic analysis code simulation Transient disturbance

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Thermal-hydraulic analysis of helical coil once-through steam generators under complex oceanic conditions

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PROGRESS IN NUCLEAR ENERGY 2024年 177卷

作者： Zhang, Chao Li, Jiangping Liu, Zihan Hou, Yandong Zhou, Yunlong Northeast Elect Power Univ Sch Energy & Power Engn Jilin Peoples R China

The helical coil once-through steam generator (HCOTSG) possesses several advantages, including excellent thermal efficiency, compact size, rapid start-up, and low energy consumption. These qualities make it particularly well-suited for use in ship reactors. This paper introduces a one-dimensional thermal-hydraulic calculation program for helical coil once-through steam generators (HCOTSGs) operating under complex oceanic conditions, utilizing the finite difference method alongside fixed boundary conditions and staggered grid techniques. By incorporating the heat transfer and friction relations to the helical coil, as well as considering the geometric structure of the helical coil when incorporating ocean conditions, the characteristics of the helical coil are reflected in the one-dimensional code. The impact of oceanic conditions is considered by modifying the momentum equation. This paper considers highly complex oceanic conditions that involve a coupling of inclining and oscillation. Furthermore, the accuracy of the developed program is verified by comparing the HCOTSG design parameters of the MRX reactor with the calculated values from internationally recognized prediction programs against the findings presented here.

关键词： Helical coil once-through steam generator Ocean condition code simulation Thermal-hydraulic

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